1. Field of the Invention
The present invention relates in general terms to pressurized water nuclear reactors and more specifically relates to the problem of the need of removing the residual power or after-power from the core in the case of a programmed or accidental reactor shutdown.
2. Brief Description of Related Art
Firstly the term residual power will be defined. On shutting down a reactor by introducing a high antireactivity into the core, the number of fissions in the latter becomes very rapidly negligible after a few seconds. However, the radioactivity of the fission products developed in the core during the normal reactor operating period continues to release a significant power, which can represent approximately 7% of the operating power at the time of reactor shutdown. Therefore, no matter why the shutdown has taken place and in particular when it occurs as a result of a depressurization incident with respect to the primary circuit, it is necessary to have means for removing said residual power or after-power from the core without the heating leading to catastrophic conditions and which could even bring about core meltdown.
Conventionally three means have been used up to now for removing the residual power from pressurized water reactors. They are constituted by the steam generator, the system for cooling the reactor on shutdown and the safety injection device for accidental situations.
The steam generator, whose normal function is to absorb heat, can obviously continue to serve a heat exchange function with the primary water following reactor shutdown. This process, which can last several hours, becomes inoperative when the pressure and temperature respectively drop to approximately 30 bars and 180.degree. C. Thus, the steam generators and secondary circuit are not designed for removing heat at low temperature and low pressure.
As from this time it is the system for cooling the reactor on shutdown which comes into action by injecting cold water into the primary circuit. Thus, within about 15 hours it is possible to bring the core to a temperature below about 100.degree. C.
The safety injection circuit ensures the emergency cooling of the core and the rapid insertion of antireactivity into it in all cases where there is an accidental depressurization of the primary circuit and which can even lead to a complete break in said circuit. It fulfils its function by as rapidly as possible injecting boric acid cooling solution into the reactor core.
These various means, whose operation is satisfactory unfortunately suffer from a number of deficiencies, which will be given hereinafter.
The distance between the cold air source and the core can lead to an inadequate operation of these means. Thus, the more equipment existing between the core and the cold source, the greater the failure risk (pipe breaks, poor operation of a valve, motor, etc.).
The design of the steam generator only enable it to operate at high pressures and temperatures. At low pressures and temperatures, the shutdown reactor cooling system is used for removing the residual power. Generally, the operational overlap range of the two systems is narrow and requires a special procedure.
During intervention on the steam generator, the water level in the primary circuit is at mid-height in the hot and cold pipes and the shutdown reactor cooling system openings are just below this level. Special precautions relative to the operation of the shutdown reactor cooling system have to be taken, so as to avoid any air entrainment risk and the formation of vortexes leading to the disappearance of the residual power removal function.
Following a primary coolant loss incident, the steam generators and shutdown reactor cooling system can become completely unavailable, even on a long term basis. The only way to remove the residual power is the safety injection device, which is an active system. However, in this hypothesis, a possible disappearance of electric sources leads to a stoppage to the removal of the residual power.
As has been shown, existing systems may be defective and this may lead to serious consequences for the reactor and its environment.
Various solutions have already been proposed for improving the safety of the nuclear reactor residual power removal apparatus. Virtually all the solutions proposed consist of introducing an auxiliary heat exchanger into the reactor vessel. Reference can be made in this connection to the CEA FR-A-8,103,632, which recommends the introduction of an exchanger into the reactor vessel for extracting the heat from the heat transfer fluid without using loops. However, in order for such a system to be effective, it is necessary for the heat transfer fluid to be able to flow between the core and the exchanger. This arrangement within the actual vessel is not described and the vessel design proposed is completely different from that of presently used vessels.
Other documents, such as the article "A water level initiated decay energy cooling system" by Charles W. Forsberg, Oak Ridge National Laboratory, pp. 229 ff, Nuclear Technology, Vol. 96, November 1991, also describe water reactors with integrated exchangers for removing residual power. These are astute "heat switch" systems controlling the heat exchange between the primary circuit and the exchanger. However, these systems are cumbersome, are not compatible with existing pressurized water reactors and are really intended for other reactor types.